IR 05000333/1986010

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Insp Rept 50-333/86-10 on 860621-0808.No Violation Noted. Major Areas Inspected:Ler Review,Operational Safety Verification & Surveillance Observations.Instruments That Control Min Flow Valves in ECCS Inadequately Tested
ML20206S763
Person / Time
Site: FitzPatrick Constellation icon.png
Issue date: 09/05/1986
From: Linville J
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20206S747 List:
References
50-333-86-10, NUDOCS 8609220365
Download: ML20206S763 (8)


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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

DCS Nos.

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50333-860525 50333-860703 Report N Docket N License No. DPR-59 Category C Licensee: Power Authority of the State of New York P.O. Box 41 Lycoming, New York 13093 Facility Name: J.A. FitzPatrick Nuclear Power Plant Inspection At: Scriba, New York Inspection Conducted: June 21 - August 8, 1986 Inspector: A.J. Lupt , Senior .esident Inspector Approved by: h Jb s Ckrinville S ' tion Chief, Reactor f

rojects Sec i 2C, DRP

/ 04te Inspection Summary:

Inspection on June 21 - August 8, 1986 (Report N /86-10)

Areas Inspected: Routine and reactive inspection during day and backshift hours by one resident inspector (153 hours0.00177 days <br />0.0425 hours <br />2.529762e-4 weeks <br />5.82165e-5 months <br />) of licensee action on previous inspection findings, licensee event report review, operational safety verification, surveillance observations, maintenance observations, followup of a plant trip and review of periodic and special report Results: During the inspection no violations were noted. However, several safety related instruments which control the minimum flow valves in Emergency Core Cooling Systems were found not to be part of a periodic calibration program and were not being adequately tested during surveillance testin (paragraph 6e.)

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DETAILS Persons Contacted -

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[ During this' inspection period, the inspector interviewed or held discus-sions with operators, technicians, and maintenance, contractor, engineer-

ing, administrative and supervisory personnel.

' Summary of Plant Activities i

The plant inspection period began with the plant operating at full power.-

On July 3,1986, the plant tripped from full power due to a failure in the

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protective relaying circuit for the 345kv output distribution line. The

! plant returned to power operation on July 4, 1986 and continued at full

power throughout the remainder of the inspection period.

! Licensee Action on Previous Inspection Findings l (Closed) Unresolved Item (77-26-03) This item identified Containment

Atmosphere Dilution (CAD) valves which were not locked as specified in the
FSAR. The inspector verified that the valves identified as being locked'

i in Figure 5.2-9 of the FSAR were also listed as being locked in the valve j lineup checklist in Operating Procedure 37, Nitrogen Ventilation and Purge, CAD, Containment Vacuum Relief and Containment Differential Pres-

sure Systems, Revision 24, dated August 15, 1985. Using the valve check-

[ list, the inspector verified the position of the locked valves. The inspector had no further questions and considers this item close (Closed) Unresolved Item (77-26-02) This item was unresolved pending the

] completion of the preoperational test of the CAD. system and demonstration

of proper annunciator logic. The inspector reviewed modification package

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FI-76-90 and verified that the preoperation test 27A-1 was completed

demonstrating proper annunciator logic. The inspector also verified by i discussions with operators and by observations, that the annunciator i circuit has re-flash capability, such'that following acknowledgement of an

! alarm at the local panel the control room annunciator will clear and j therefore be available in the event of subsequent alarms.

' Licensee Event Report (LER) Review

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i The Inspector reviewed LERs to verify that the details of the events were clearly reported. The inspector determined that reporting requirements

! had been met, the report was adequate to assess the event, the cause l appeared accurate and was supported by details, corrective actions l appeared appropriate to correct the cause, the form was complete, and l- generic applicability to other plants was not in question.

During this inspection period the following LERs were reviewed.

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LER 86-12 reported that the High Pressure Coolant Injection System was inoperable. Details of -his event are discussed in paragraph 5 of Inspec-tion No. 50-333/86-0 LER 86-13 reported a reactor trip due to a main turbine trip. The turbine trip was caused by a fault in a protective relaying circuit in the 345ky distribution system. Details of this event are discussed in paragraph . Emergency Notification System Reports The inspector reviewed the following events which were reported to the NRC via the Emergency Notification System as required by 10 CFR 50.72. The review included a determination that the reporting requirements were met, that appropriate corrective actions have been taken, and the event was evaluated for possible generic implication The following reports were reviewed:

Event Date _ Subject July 3, 1986 Reactor trip due to failure in the protective relaying circuit of the 345kv output distribution system. This event is also discussed in paragraph . Operational Safety Verification Control Room Observations Daily, the inspector verified selected plant parameters and equipment availability to ensure compliance with limiting conditions for operation of the plant Technical Specifications. Selected lit annunciators were discussed with control room operators to verify that the reasons for them were understood and corrective action, if required, was being taken. The inspector observed shift turnovers bi-weekly to ensure proper control room and shift manning. The inspector directly observed the operations listed below to ensure adherence to approved procedures:

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Routine power operation Issuance of RWP's and Work Requests / Event /

Deficiency form No violations were identifie Shift Logs and Operating Records Selected shift logs and operating records were reviewed to obtain information on plar.t problems and operations, detect changes and trends in performance, detect possible confitets with Technical Specifications or regulatory requirements, determine that records are beirg maintained and reviewed as required, and assess the effective-ness of the communications provided by the log ..

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No violations were identifie Plant Tours During the inspection period, the inspector made observations and-conducted tours of the plant. During the plant tours, the inspector conducted a visual' inspection of selected piping between containment and the isolation valves for leakage or leakage paths. This included verification that manual valves were shut, capped and locked when required and that motor operated valves were not mechanically blocked. The inspector also checked fire protection, housekeeping /

cleanliness, radiation protection, and physical security conditions to ensure compliance with plant procedures and regulatory require-ment No violations were identifie Tagout Verification The inspector verified that the following safety-related protective tagout records (PTR's) were proper by observing the positions of breakers, switches and/or valves:

-- PTR 861083 on "B" Residual Heat Removal Syste PTR 861026 on "C" Residual Heat Removal Service Water Syste PTR 861123 on "B" Standby Gas Treatment Syste ,

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No violations were identifie Emergency System Operability The inspector verified operability of the following systems by ensur-ing that each accessible valve in the primary flow path was in the correct position, by confirming that power supplies and breakers were properly aligned for components that must activate upon an initiation signal, and by visual inspection of the major components for leakage and other conditions which might' prevent fulfillment of their functional requirements:

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Reactor Core Isolation Cooling Syste Containment Atmosphere Dilution System.

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Fire Protection Water Syste "A" Low Pressure Coolant Injection Independent Power Supply Syste , a

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-While conducting a walkdown of the Reactor Core Isolation Cooling:

(RCIC) System, the inspector checked system instrumentation for proper calibration. The inspector determined that pressure switch 13-PS-127 and flow switch 13-FS-57, used to control the minimum flow valve for the _RCIC system, were not part of any routine calibration program. Further inspection determined that instrumentation which controls _the minimum flow valves for the High Pressure Coolant Injection System (HPCI), Residual Heat Removal (RHR) System, and the Core Spray System were also not in a periodic calibration progra These minimum flow valves automatically open on low flow and shut on high flow to protect the pumps from overheating when operating with their discharge valves shut. Section 7.4.4 of the licensee's Final

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Safety Analysis Report lists the minimum flow valves as an. example of a device that could interrupt planned Emergency Core Cooling System operation and must act to prevent complete failure of the component or syste The operability of the minimum flow valves are checked during Oper-ations Department Surveillance Tests (STs); however, several defic-iencies were noted. In the ST for the Core Spray System, the set-point _at which the' valve is required to operate cannot be accurately measured with the instrumentation available to the operator. This is also the case for the RHR Syste In addition, the RHR System ST does not give a specific flow rate at which the valve should operat The licensee will add these flow switches to their routine calibration program and plan to calibrate them during a short outage scheduled in September. Although the STs for the HPCI and RCIC Systems check that the minimum flow valves shut at a required flow'and the instrumentation is sufficient to monitor this flow, the STs do 'not verify either the flow at which the valves open or that they did open. The licensee ;

will review the STs to ensure that they adequately verify the oper- '

ability of the minimum flow valve Based on these findings, the inspector asked if other safety related instrumentation existed which was not under a periodic calibration or testing program. In a limited review, the licensee identified approximately 400' instruments, classified as Quality ~ Assurance (QA)

Category 1, which are not under any periodic calibration program. It must be noted that some of these instruments are classified as Cate-gory 1 for reasons other than performing a safety function (eg. part of a pressure boundary) and others may be tested under other surveil-lance procedures. Also, the licensee is currently preparing a pro-gram to generate a new QA Category list as part of a Master Equipment List (MEL). The MEL will ensure that the proper classification is assigned to each componen The licensee stated that they will review these 400 instruments within 6 months to determine whether they should be added to the periodic calibration program. A more detailed determination of testing and calibration requirements for all instruments is scheduled to be conducted following completion of the MEL, as part of the

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11censee's Planned Maintenance Upgrade Program. However, this review is not expected to be completed for the. safety related instruments until late 1987. The inspector will review the licensee's response and corrective actions to these concerns during subsequent inspections. (UNR86-10-01) Surveillance Observations The inspector observed portions of the surveillance procedures listed below to verify that the test instrumentation was properly calibrated, approved procedures were used, the work was performed by qualified person-nel, limiting conditions for operation were met, and the system was

correctly restored following the testin F-ST-2R, Residual Heat Removal Service Water Pump and Motor Operated Valve Operability Test, Revision 13, dated April 9, 1986, performed June 26 and July 14, 198 F-ISP-175A, Reactor and Containment Cooling Instrument Functional Test / Calibration, Revision 4, dated May 14, 1986, performed July 3, 198 F-ST-2A, Residual Heat Removal Pump Flow Rate Test, Revision 18, dated March 26, 1986, performed June 26, 198 F-ISP-19, Off-Gas Radiation Monitor Instrument Calibration, Revision 16, dated July 17, 1986, performed July 24, 198 The inspector also witnessed all aspects of the following surveillance test to verify that the surveillance procedure conformed to technical specification requirements and had been properly approved, limiting condi-tions for operation for removing equipment from service were met, testing was performed by qualified personnel, test results met technical specifi-cation requirements, the surveillance test documentation was reviewed, and equipment was properly restored to service following the tes F-ISP-70, Reactor High Pressure Permissive Instrument Functional Test / Calibration, Revision 11, dated May 14, 1986, performed August 6, 198 i No violations were identifie . Maintenance Observations The inspector observed portions of various safety-related maintenance activities to determine that redundant components were operable, that

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these activities did not violate the limiting conditions for opera-tion, that required administrative approvals and tagouts were obtained prior to initiating the work, that approved procedures were used or the activity was within the " skills of the trade," that appropriate radiological controls were properly implemented, that

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ignition / fire prevention controls were properly implemented, and that equipment was properly tested prior to returning it to servic During this inspection period, the following activities were observed:

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WR 02/43200, replace level transmitter 728 which initiates Emergency Core Cooling System WR 10/42638, remove and inspect "C" Residual Heat Removal Service Water Pum WR 71/31222, replace test blocks in the 345ky switchyar PMWR 10/03243, perform preventive maintenance on the limitorque valve operator for the "D" Residual Heat Removal Pump Shutdown Cooling Suction Valv No violations were identifie . Followup of Plant Trip At 12:20 a.m. on July 3, 1986, the reactor tripped from full power due to a main turbine trip. The turbine trip was caused by a protective relay sensing an erroneous high differential current between the 345kv output lines. The high differential current signal was due to a failure of a test block in the protective relaying circuit. The test block can be used to isolate the relaying circuit for testing while the unit is on-lin There was no Emergency Core Cooling System actuation or any radioactive release associated with this even The inspector reviewed the process computer alarm printout, the post-trip log, various chart recorders, and the completed data sheets for procedure PS0 No. 53, " Post Trip Evaluation". Based on these reviews, the inspector determined that the operator actions in response to the event were proper and in accordance with approved procedures, the plant responded as designed, and the licensee review and determination of root cause was

, adequat The failure of.the test block was caused by cracking and separation of one

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of the terminal studs. Stud failures in the test blocks had previously been found at other f acilities and IE Information Notice No. 85-83, dated October 1985, was issued to alert the industr In reponse to this notice, the licensee had checked two test blocks used

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in the safety related recirculation pump motor generator circuit, but was

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unaware that test blocks existed in the non-safety related electrical distribution system. In a Service Advice Letter dated May 28, 1986,

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General Electric identified the cause of the previous stud failures as stress corrosion cracking due to use of an improper alloy, which had not been stress relief annealed as required.

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The licensee checked other test blocks in the 345kv switchyard protective relaying circuit by applying five pounds of force to the studs as des-cribed in the IE Notice. One additional failure occured during this testing. Both test blocks were replaced prior to startup, with blocks constructed of the proper material. Other test blocks have been identified in the 115kv protective relaying circuitry. A method of safely testing the 115kv test blocks and identification of any other test blocks will be completed by December 1986. The inspector will review these licensee actions during a subsequent inspection (333/85-10-02).

10. Review of Periodic and Special Reports Upon receipt, the inspector reviewed periodic and special reports. The review included the following: inclusion of information required by the NRC; test results and/or supporting information consistent with design predictions and performance specifications; planned corrective action for resolution of problems, and reportability and validity of report infor-mation. The following periodic report was reviewed:

-- June 1986 Operating Status Report, dated July 10, 1986 '

1 Exit Interview At periodic intervals during the course of this inspection, meetings were held with senior facility management to discuss inspection scope and find-ings. On August 12, 1986, the inspector met with licensee representatives and summarized the scope and findings of the inspection as they are described in this repor Based on the NRC Region I review of this report and discussions held with licensee representatives during the exit meeting, it was determined that this report does not contain information subject to 10 CFR 2.790 restric-tions.